ML13009A299
| ML13009A299 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 02/13/2013 |
| From: | Joel Wiebe Plant Licensing Branch III |
| To: | Pacilio J Exelon Generation Co |
| Joel Wiebe, NRR/DORL 415-6606 | |
| References | |
| TAC ME8470, TAC ME8471 | |
| Download: ML13009A299 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 13, 2013 Mr. Michael J. Pacilio Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)
Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNIT 2, AND BYRON STATION, UNIT NO.2, CLOSURE EVALUATION FOR REPORT PURSUANT TO TITLE 10 OF THE CODE OF FEDERAL REGULATIONS, PART 50, SECTION 50.46, PARAGRAPH (a)(3)(ii) CONCERNING SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERROR RELATED TO NUCLEAR FUEL THERMAL CONDUCTIVITY DEGRADATION (TAC NOS. ME8470 AND ME8471)
Dear Mr. Pacilio:
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46, paragraph (a)(3)(ii), Exelon Generation Company, LLC, (the licensee), submitted a report describing a significant error identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effect of the error on the predicted peak cladding temperature.
The report was submitted by letter dated March 19, 2012. The licensee submitted a response to a request for additional information by letter dated June 11, 2012.
The U.S. Nuclear Regulatory Commission staff evaluated the report, as supplemented, and has determined that the report satisfies the intent of the reporting requirements promulgated at 10 CFR 50.46(a)(3)(ii), as discussed in the statement of considerations published in the Federal Register (FR) on September 16, 1988 (53 FR 35996), for the realistic ECCS evaluations revision of 10 CFR 50.46.
The staff evaluation is enclosed.
M. Pacilio
- 2 Is you have any questions, please contact me at 301-415-6606.
Sincerely,
~~~~ject Manager a~~nt Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-457 and STN 50-455
Enclosure:
Staff Evaluation cc w/encl: Distribution via Listserv
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- i< CLOSURE EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION, UNIT 2, AND BYRON STATION, UNIT NO.2 REPORT DESCRIBING THE NATURE OF AND ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE OF A SIGNIFICANT EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL ERROR DOCKET NOS 50-457 and 50-455
1.0 INTRODUCTION
By letter dated March 19, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12079A112), Exelon Generation Company, LLC, (Exelon, the licensee) submitted a report describing a significant error identified in the emergency core cooling system (ECCS) evaluation model, and an estimate of the effect of the error on the predicted peak cladding temperature (PCT) for Braidwood Station, Unit 2, and Byron Station, Unit No.2. This report was submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.46, paragraph (a)(3)(ii). The report was supplemented by letter dated June 11,2012 (ADAMS Accession No. ML121640840), and referred to a letter from Westinghouse Electric Company dated March 7, 2012 (ADAMS Accession No. ML12072A035).
Braidwood Station, Unit 1, and Byron Station, Unit No.1 will be addressed in a separate safety evaluation (SE).
The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff has evaluated the report, along with its supplemental information, and determined that it satisfies the reporting requirements of 10 CFR 50.46(a)(3), and also the intent of the reporting requirements, as discussed in the statement of considerations published in the Federal Register (FR) on September 16, 1988 (53 FR 35996), for the realistic ECCS evaluations revision of 10 CFR 50.46. The staff review is discussed in the following sections.
2.0 REGULATORY EVALUATION
2.1 Requirements Contained in 10 CFR 50.46 Acceptance criteria for ECCSs for light-water nuclear power reactors are promulgated in 10 CFR 50.46. In particular, 10 CFR 50.46(a)(3)(i) requires licensees to estimate the effect of any change to, or error in, an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For the purpose of 10 CFR 50.46, a significant change or error is one which results in a calculated peak fuel cladding temperature different by Enclosure
-2 more than 50 degrees Fahrenheit (OF) from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 OF.
For each change or error discovered in an acceptable evaluation model, or in the application of such a model, 10 CFR 50.46(a)(3)(ii) requires the affected licensee to report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually. If the change or error is significant, the licensee is required to provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements.
2.2 Additional Guidance Additional clarification concerning the intent of the reporting requirements is discussed in the statement of considerations for the realistic ECCS evaluation revision of 10 CFR 50.46 (53 FR 35996):
[Paragraph (a)(3) of section 50.46] requires that all changes or errors in approved evaluation models be reported at least annually and does not require any further action by the licensee until the error is reported. Thereafter, although reanalysis is not required solely because of such minor error, any subsequent calculated evaluation of ECCS performance requires use of a model with such error, and any prior errors, corrected. The NRC needs to be apprised of even minor errors or changes in order to ensure that they agree with the applicant's or licensee's assessment of the significance of the error or change and to maintain cognizance of modifications made subsequent to NRC review of the evaluation modeL..
Significant errors require more timely attention since they may be important to the safe operation of the plant and raise questions as to the adequacy of the overall evaluation model... More timely reporting (30 days) is required for significant errors or changes... this final rule revision also allows the NRC to determine the schedule for reanalysis based on the importance to safety relative to other applicant or licensee requirements.
The NRC staff considered this discussion in the FR in its evaluation of the error report submitted by the licensee.
3.0 TECHNICAL EVALUATION
In Exelon's March 19,2012, letter, the licensee responded to an information request from the NRC regarding the effects of Thermal Conductivity Degradation (TCD) on PCT. The licensee estimated that the effects of TCD would have a +148 OF difference in PCT from the analysis of record for both Braidwood, Unit 2 and Byron, Unit No.2. The licensee reported their results via the 10 CFR 50.46 reporting requirements. The staff evaluation is based on a graded
-3 approach. In this approach, the licensees who had an abnormal occurrence report PCT greater than 2000 of were sent a request from the NRC.
The report submitted by the licensee described the effects of an error in the ECCS evaluation model associated with the degradation of thermal conductivity in nuclear fuel. This issue is discussed in NRC Information Notice (IN) 2009-23, "Nuclear Fuel Thermal Conductivity Degradation." Its potential effects in realistic ECCS evaluation models are described in IN 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation."
Based on the nature of the reported error, and on the magnitude of its effect on the PCT calculation, the NRC staff determined that a more detailed technical review was necessary.
The NRC staff's review was performed to ensure that it agrees with the licensee's assessment of the significance of the error, and to enable the NRC staff to verify that the evaluation model, as a whole, remains adequate. Finally, the NRC staff's review also establishes that the licensee's proposed schedule for re-analysis is acceptable in light of the safety significance of the reported error and the proposed steps taken to ensure that there is adequate safety margin to ensure compliance with 10 CFR 50.46(b) acceptance criteria.
3.1 Overview of Automated Statistical Treatment of Uncertainty Method The licensee uses the NRC-approved Automated Statistical Treatment of Uncertainty Method (ASTRUM), documented in WCAP-16009-NP-A (ADAMS Accession Nos. ML050910157, ML050910159, and ML050910161, respectively), to evaluate ECCS performance. ASTRUM relies on an approach based on order statistics, in which a set number of cases with randomly varied initialconditions are analyzed using the WCOBRAITRAC (WcrT) reactor system analysis code. Thenumber of cases is chosen so that the highest predicted PCT within the case set becomes a predictor of the 95/95 upper tolerance limit for the PCT associated with a hypothetical population of loss-of-coolant accident (LOCA) scenarios. This result is used to show compliance with the 10 CFR 50.46(b)(1) acceptance criterion concerning PCT.
3.2 Summary of Technical Information in the Report The licensee's report indicated that the effect of the TCD error was +148 of. The nature of the error, and the method used to estimate its effect on the calculated PCT, is discussed in greater detail in the March 7,2012, Westinghouse letter and the request for additional information (RAI) responses. In the report, the licensee also discussed additional changes made to the ECCS evaluation in order to offset the effects of TCD, and to recapture margin to the regulatory limit on PCT.
TCD Error Correction The error in the ECCS evaluation model was caused by the inability of the Fuel Rod Performance and Design (PAD) fuel performance model to account for the effects of TCD with increasing fuel burnup. This error caused fuel temperature initial conditions to be non conservatively low for higher burnup fuel rods that were analyzed in the ECCS evaluation. In order to correct for the error, a burnup-dependent term was added to the nuclear fuel thermal
-4 conductivity equation which caused the predicted initial fuel temperatures [
The licensee used PAD 4.0 + TCD to generate inputs to the ASTRUM execution, which is different from the existing analysis. In the typical ASTRUM analysis, the WCIT initialization generates the fuel conditions using a MATPRO-based1 analytic procedure. The initial conditions are also calculated using PAD 4.0, and then the WCIT fuel temperature is corrected to the PAD fuel temperature by adjusting fuel rod plenum heat transfer properties. With the improved PAD correction, the [
] is used for the initialization, instead. The
[
] more closely simulates the fuel performance predicted by PAD 4.0 + TCD than the MA TPRO model.
In order to estimate the PCT effect of the TCD error correction, the licensee [
Additional Model Changes to Offset TCD Effects In order to ensure that facility operation remains compliant with 10 CFR 50.46 requirements and to restore margin, the licensee made additional model changes to offset the increase in predicted PCT due to the TCD model error.
The licensee made three model changes in order to offset the increase in predicted PCT due to the TCD model error. The changes included the reduction of the upper bound steam generator tube plugging from 10 percent to 5 percent reduction in the upper bound nominal average vessel temperature from 588 OF to 583.5 OF, and an increase in the conservatively low containment pressure boundary. All of the changes made were to the model input and were not operating changes. The changes resulted in recapture of analytic safety margin.
['
2Hagrman, D.L., G.A. Reymann, and G.E. Mason. 1981. A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, MATPRO Version 11 (Revision 2), NUREG/CR 0479 (TREE-1280), prepared by EG&G Idaho, Inc., Idaho Falls, 10, for the U.S Nuclear Regulatory Commission, Washington, D.C.
[3
-5 The net effect of incorporating all of the model changes, including TCD, was a reduction in predicted PCT of 190 OF.
Reported Results Following the correction for TCD and the additional ECCS model changes, the current predicted PCT for Braidwood, Unit 2, and Byron, Unit No.2, is 1999.0 of. The licensee has also provided a commitment to perform a reanalysis using an NRC-approved evaluation model that accurately considers TCD, when such a model becomes available. This submittal is currently expected in 2016.
3.3 Summary of Staff Evaluation The NRC staff evaluation of the error report submitted by the licensee included a review of the report itself, a detailed audit to review the analyses supporting the report, and an RAI, to which the licensee responded via the June 11, 2012, letter. The staff performed a detailed review of the input parameters and limitin results that were used to generate the estimate, and concluded that the estimate enables the current analysis to maintain a high level of probability that the 2200 of PCT acceptance criterion is not exceeded.
The NRC staff conducted an audit at the Westinghouse facilities which consisted of reviewing the licensee's evaluation of the effects of TCD on PCT. As a result of the audit, the NRC staff determined the need to request information pertaining to the modeling changes implemented to access the PCT. The first request to the licensee was for the data used as the ASTRUM input in the analysis of record. The NRC staff compared this input data in its entirety to the input data for the estimate in order to determine if the bounding and most limiting cases were included in the estimate analysis. It was concluded that data used was bounding. The licensee and their vendor explained how and why data was included or excluded from the estimate.
The second request by the NRC staff was for a justification of containment pressure changes.
In order to obtain analytic margin, the licensee changed the containment pressure of the system to the newly calculated containment pressure that resulted when the steam generator tube plugging limit and reactor vessel temperature was reduced. All of the above listed changes were within the as-operated conditions and resulted in a more realistic estimate.
The NRC staff asked about any changes to fuel characteristics that were made in the licensee's estimate. There were no changes made to the fuel characteristics in the estimate because the estimate used bounding fuel inputs that are used across the Exelon fleet. The fuel used in the analysis had Exelon specific fuel rod dimensions, plant operating parameters, and rod average linear heat generation using an Exelon specific bounding power history.
The NRC staff requested information about the specific equations used to calculate TCD and if the equations were consistent between each computer code used in the analysis. The licensee stated that the evaluation did not include any error corrections, code improvements, or model changes from the analysis-of-record code versions. The licensee also evaluated the differences between the HOTSPOT and PAD thermal conductivity models. The thermal conductivity
-6 degradation model results were [
]. The licensee concluded that for a given maximum fuel average temperature and burnup, the differences between the PAD 4.0 TCD and WCOBRAITRAC and HOTSPOT fuel thermal conductivity models [
]. The licensee provided the staff with information which supported this conclusion.
Additionally, the NRC staff asked the licensee to provide additional detail concerning the steady state ASTRUM initialization process. The staff needed clarification as to what fuel characteristics are adjusted within the applicable models to obtain convergence among HOTSPOT, WCOBRA-TRAC, and PAD4.0 TCD. Exelon listed the [
] used to determine steady state. Of these [
]. The licensee provided the NRC staff with the necessary information to determine that they used an acceptable initialization process.
The NRC staff requested information pertaining to the processes used by the licensee and vendor to ensure the LOCA analysis input variables conservatively bound the as-operated plant conditions. The licensee explained that the input variables are determined in the fuel reload process. The licensee states, "the generic fuel reload evaluation approach relies upon the bounding approach in which safety analyses are performed to accommodate the plant changes resulting from different or new fuel in the core without requiring new safety analyses." In the process, important parameters in LOCA analysis which show a fuel reload dependency are generated, evaluated, and confirmed to support the reload. Finally, additional plant changes are evaluated by the licensee and their vendor to determine their applicability to the LOCA analysis each cycle. The staff finds this to be an acceptable approach to ensure the LOCA analysis input variables conservatively bound the as-operated conditions.
The licensee provided a commitment to perform a re-analysis, stated as follows:
EGC will submit for review and approval a LBLOCA [large-break loss-of-coolant accdent]
analysis that applies NRC approved methods that include the effects of fuel TCD for Braidwood Station, Unit 2 and Byron Station, Unit 2. The date for the analysis submittal is contingent on the following milestones which must be completed in order to perform a revised licensing basis LBLOCA analysis with an NRC approved EGGS evaluation model that explicitly accounts for TCD:
- a. NRC approval of a fuel performance analysis methodology that includes the effects of TCD. The new methodology for developing inputs to the LBLOCA evaluation model would replace the current licensing basis methodology for Braidwood Station, Unit 2 and Byron Station, Unit 2 that is described in WCAP 15063-P-A, Revision 1, with Errata, 'Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)."
- b. NRC approval of a LBLOCA evaluation model that includes the effects of TCD and accommodates the rulemaking associated with the proposed 10 CFR 50.46c (Docket 10 NRC-2008-0332). The new methodology would replace the current licensing basis methodology, WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."
- 7 Since the licensee's evaluation is based on a very rigorous analytic approach, the NRC staff finds that the licensee has demonstrated that the effects of the error are appropriately estimated, and that the licensee has provided assurance that it will not exceed 2200 OF following a LOCA The NRC staff may find that it is necessary to revisit this conclusion if other significant errors in the ASTRUM evaluation model are reported.
The re-analysis requirement contained in 10 CFR 50.46(a)(3)(ii) states the following:
... and [the licensee] shall include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with [10 CFRJ 50.46 requirements.
The commitment provided by the licensee satisfies this requirement by indicating a proposed date that a re-analysis will be provided. Further, the NRC staff finds that the proposed re-analysis date is commensurate with the safety significance of the issue, based on the considerations described above. Therefore, the NRC staff finds that the licensee has adequately addressed the reanalysis requirement contained in 10 CFR 50.46(a)(3)(ii).
4.0 CONCLUSION
Based on the considerations discussed above, the NRC staff finds that the report for Braidwood Station, Unit 2, and Byron Station, Unit No.2, submitted pursuant to 10 CFR 50.46(a)(3)(ii),
concerning an ECCS evaluation model error pertaining to TCD, satisfies the intent of the 10 CFR 50.46 reporting requirements. The submittals dated March 19,2012, and June 11, 2012, enabled the staff to (1) determine that it agrees with the licensee's assessment of the significance of the error, (2) confirm that the evaluation model remains adequate, (3) verify that the licensee continues to meet the PCT acceptance criterion promulgated by 10 CFR 50.46(b),
and (4) determine that the licensee's proposed schedule for re-analysis is acceptable in light of the information provided. An evaluation for Braidwood Station, Unit 1 and Byron Station, Unit No.1 will be completed in a separate safety evaluation report.
M. Pacilio
- 2 Is you have any questions, please contact me at 301-415-6606.
Sincerely, IRAI Joel S. Wiebe, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-457 and STN 50-455
Enclosure:
Staff Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
EUBLIC LPL3-2 RIF RidsNrrPMByron Resource RidsNrrPMBraidwood Resource RidsNrrLASRohrer Resource RidsNrrDorlLpl3-2 Resource RidsOgcRp Resource RidsRgn3MailCenter Resource RidsNrrDorlDpr Resource RidsAcrsAcnw_MailCTR Resource ADAMS Accession No ML13009A299 OFFICE NAME DATE LPL3-2/PM JWiebe 2/12/13 LPL3-2/LA SRohrer 2/12/13
! LPL3-2/BC NDiFrancesco (A) 2/13/13 LPL3-2/PM JWiebe 2/13/13 OFFICIAL RECORD COpy